Canadian Nuclear Laboratories Research Articles

Pilot Scale Validation of a Chemical Process for Uranium, Cesium, and Mercury Recovery from Cemented Radioactive Wastes

The medical isotope (Mo-99) production at Chalk River Laboratory involves the dissolution of irradiated isotope targets prior to the extraction of Mo-99. This process generates a waste that is cemented in 5-gallon containers and transferred to a waste-management facility for intermediate storage. Over the past decades, a large number of five-gallon containers of cemented radioactive waste (CRW) were produced, and Canadian Nuclear Laboratories (CNL) must develop a process to convert this material to a permanent waste form. Research has been undertaken to develop an innovative method for the recovery of U, Hg, and Cs from surrogate cemented radioactive waste (SCRW). This paper presents the pilot scale validation of the operating parameters prior to the demonstration scale testing. Leaching tests at the pilot scale were conducted with 5 kg of SCRW to validate the main operating parameters and evaluate the reuse of the leaching solution. The mean solubilization yields obtained at the pilot scale were 81.9 ± 8.3% for Cs, 99.0 ± 1% for U, and 94.9 ± 4.5% for Hg. Columns with 100 g of KNiFC-PAN and 250 g of Lewatit TP214 allow for the separation of Cs and Hg from 60 L of leaching solution without U loss. Flow rates of 12.5 BV/h and 25 BV/h were suitable to achieve 99% separation of Hg and Cs, respectively. For the Hg resin, the capacity reached 23.4 mg/g, and the capacity for the Cs resin reached 0.79 mg/g. The pilot scale U extraction results showed that the U adsorption is selective, with a breakthrough at 36 BV (capacity for U of 3.70 mg/g). Uranium elution with 1 M Na2CO3 exceeded 99%, and subsequent precipitation with NaOH achieved 99% recovery. SEM data confirmed the high purity of the U solids produced as sodium di-uranate.

The impact of 40 years of radiation on the integrity of copper

Cu piping previously used to carry process air was harvested from the now shutdown National Research Universal reactor at the Canadian Nuclear Laboratories site in Chalk River, Ontario. Metallurgical and microscopy examinations found no quantifiable differences between irradiated and non-irradiated sections of the pipe. This provides evidence for the safety case for the use of the Cu-coated used fuel container proposed by the Nuclear Waste Management Organization for the permanent disposal of high-level nuclear waste.

Investigation of Critical Heat Flux for Plutonium-Based Mixed Oxide Advanced Fuel Bundle Design

Abstract The critical heat flux (CHF) performance of an advanced plutonium-based mixed oxide (MOX) fuel for potential use in a pressure tube heavy water reactor (PT-HWR) has been studied experimentally at Canadian Nuclear Laboratories with an electrically heated string simulator of 43-element fuel bundles. The fuel simulator has a uniform axial power profile and a radial power profile that is representative of the plutonium-based MOX fuel. The measurements of CHF caused by liquid-film dryout were made in the MR-3 heat transfer loop facility using R-134a refrigerant as the working fluid. The test matrix included system pressures from 1.47 to 2.11 MPa, mass flow rates from 12.7 to 14.7 kg/s and inlet temperatures from 31 to 59 °C, which are representative of the water-equivalent reactor operating conditions of 9 to 12.5 MPa pressure, 13.5 to 21.3 kg/s mass flow rate and the desired inlet subcoolings. Compared to conventional natural uranium fuel, the radial power profile of a MOX fuel exhibits a steeper and less even distribution across the fuel element rings, with a relatively higher power in the outer ring. It was found that CHF values of the MOX fuel are significantly lower than those of the natural uranium fuel. Based on the experimental data, a correlation has been derived to account for the effect of radial power profile on CHF. This correlation can be used to evaluate the relative CHF values of advanced or nonconventional fuel designs with radial power profiles deviating from that of natural uranium fuel.

Pressure drop experiment with externally and internally cooled annular fuel bundles

Canadian Nuclear Laboratories (CNL) has developed an externally and internally-cooled annular fuel (ICAF) bundle that is considered to have improved thermalhydraulic characteristics as compared to the fuel bundles used in the conventional pressurized heavy-water-moderated-and-cooled reactors (PHWRs), such as the CANDU® reactors. The ICAF bundle consists of 24 annular fuel elements (or rods) and a solid central rod. The coolant flows through both the internal and external channels of each annular fuel element; the total flow in the flow tube is split into the flow inside the internal channels of the annular elements and that in the outside subchannels of the fuel bundle. The most compelling argument for this design, compared to the conventional solid-rod design, is the significant reduction in maximum fuel temperature for equivalent linear element ratings, due to significantly-enhanced heat transfer from fuel surfaces to coolant. The primary objective of the ICAF bundle development is to achieve higher burnups and higher power densities using advanced fuel cycles, with comparable or improved thermalhydraulic safety margins. This paper presents the results of a pressure drop experiment performed to investigate the hydraulic behavior of the ICAF bundle. Experimental data were obtained for the pressure distribution, the bundle misalignment junction signature and the flow split. The data can be used to develop predictive models to be implemented into the thermalhydraulics analysis toolset, and to validate the toolset for safety analyses in establishing reactor operating power and evaluating safety margin.

Post-irradiation examination of U-7Mo/Mg and U-10Mo/Mg dispersion fuels irradiated in the NRU reactor

As part of a global initiative to reduce the usage of highly enriched uranium (HEU) fuel for research and test reactors, Al-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements with a low-enriched uranium (U-235 enrichment of 19.75 wt%) loading of 4.5 g U/cm3 were fabricated at the Canadian Nuclear Laboratories (CNL). The mini-elements were irradiated in the National Research Universal (NRU) reactor at maximum linear power ratings of 100 kW/m up to 80 at% U-235 burnup. Hot cell macroscopic examination of the discharged fuel elements at 30, 60 and 80 at% U-235 burnup indicated that they were intact without breakaway swelling or fuel cladding failure. The results of interim post-irradiation examination (PIE) that was conducted on mini-elements discharged at 30 at% U-235 burnup are reported in this paper. A follow-up paper will present the PIE results of the mini-elements that achieved burnups of 60 and 80 at% U-235. In the present study, the microstructure and phase composition of the irradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were investigated using optical microscopy and neutron diffraction analysis. The weight fraction of different crystalline phases in U-7Mo/Mg and U-10Mo/Mg fuel cores at low burnup was determined through Rietveld refinement. This paper includes a description of the irradiation tests along with preliminary PIE results from the irradiated U-Mo/Mg mini-elements and describes the results of optical microscopy characterization and neutron diffraction analysis.

A pilot study of radiation-induced bystander effect in radio-adapting frogs at a radiologically contaminated site located on the chalk river laboratories property

Purpose To measure medium borne bystander effects, to study the influence of radioadaptive response (RAR) on bystander response, and to discover reliable radioresponsive biomarkers in radio-adapting frogs from Duke Swamp contaminated with an above-background radiation level and in naïve frogs from Twin Lake as the background control site. Materials and methods Frogs were captured at Duke Swamp and Twin Lake and brought to the lab at the Canadian Nuclear Laboratories facility. Half of the frogs from each site were irradiated with 4 Gy while the other half of the frogs were left with no further radiation treatment. Frog bladders were removed and placed in sterile culture media. Upon arrival at McMaster University, the bladders were processed for tissue cultures. After 48 h, the culture media conditioned by the bladder explants were harvested for clonogenic reporter survival assay and calcium flux measurements for assessing bystander effects. HPV-G cells were used as bystander reporter cells in all radiation-induced bystander effect (RIBE) assays. The frog bladder cultures were incubated for another 10–12 days followed by immunochemical staining for bcl-2 and c-myc expressions to analyze cellular anti-apoptotic (pro-survival) and pro-apoptotic (pro-death) responses, respectively. Results Only culture media conditioned by bladders from 4-Gy-irradiated naïve frogs from Twin Lake induced bystander effects (reduction of HPV-G reporter cells’ clonogenic survival and presence of strong calcium flux activities). The 4 Gy irradiation dose increased pro-apoptotic c-myc expression in naïve frogs’ bladder explants. Culture media conditioned by bladders from radio-adapting frogs from Duke Swamp enhanced HPV-G’s clonogenic survival and a 4 Gy irradiation challenge did not change the enhanced clonogenic survival nature nor induce calcium flux. In bladder explants from both control and 4-Gy-irradiated radio-adapting frogs, anti-apoptotic bcl-2 expression for pro-survival responses was ubiquitous while c-myc expression for pro-death responses was limited to a small fraction of cells. Conclusion The clonogenic RIBE reporter assay using HPV-G and calcium flux measurements are useful diagnostic tools for RIBE assessment of field biological samples, specifically those from frogs. RAR induced by environmentally relevant low-dose radiation induces protective bystander response. Bcl-2 and c-myc are reliable biomarkers for evaluating low dose radiation responses in wild populations of amphibians. Overall, this pilot study emphasizes the importance of looking at non-targeted effects (NTEs) in natural populations of non-human biota that could be vulnerable to chronic low-dose radiation exposures.

Electric vehicle viability: evaluated for a Canadian subarctic region company

Governments all over the world are setting policies and regulations that encourage companies to incorporate Battery Electric Vehicles (BEVs) as a part of their fleet and Canada is no exception. However, BEVs struggle to deliver on a wide variety of workload needs and ambient temperature changes tend to have a greater impact on their range. This study reviewed the impact of ambient temperature on BEVs and evaluated if such depressed vehicle battery capacity can deliver required vehicle work scope for the Canadian Nuclear Laboratories (CNL) Waste Management Areas (WMA). This study focused on three-years quantitative data of Light Duty Vehicles (LDVs) operated by CNL WMA. A Total Cost of Ownership (TCO) evaluation compared BEVs to representative WMA Internal Combustion Engine (ICE) vehicles. Findings indicate that WMA LDVs mileage needs are well within the capabilities of a 150 km capacity BEV. When ambient temperature changes to 0 °C and further to −15 °C, it impacts battery capacity, reducing the aforementioned 150 km capacity BEV by 53% and 40%, respectively. Comparatively, a 250 km capacity BEV would also not translate into meeting 100% of WMA vehicle work scope without recharging when temperature impacts are considered. TCO evaluation shows cost savings in implementing BEVs over WMA ICE LDV, with or without government rebates. There is potential for BEV technology to meet WMA Small Engine Vehicles (SEVs) i.e. all-terrain vehicles (ATVs) and Utility Task Vehicles (UTVs) mileage needs without modifications, contingent on ambient temperature above −15 °C.

Demonstration of a coupled computational fluid dynamics approach for modelling non-water cooled small modular reactors

This paper provides an overview of the capabilities of a new toolset that is being developed at Canadian Nuclear Laboratories for modelling advanced non-water cooled small modular reactor concepts. Throughout the study, a technology agnostic approach was adopted. The integrated toolset includes computational fluid dynamics, neutronics and system thermalhydraulics codes to handle a range of multi-physics capabilities. The toolset has been applied to two different non-water cooled SMR concepts: molten salt and prismatic gas-cooled reactors. A step-wise approach was used in which the models were first tested using standalone codes, followed by execution of coupled-code simulations. This paper demonstrates the application of the toolset to gas-cooled and molten-salt reactor concepts, including transient simulations using the CFD code STAR-CCM+ coupled with the Monte Carlo neutronics code Serpent 2 and with the system thermalhydraulics code RELAP5-3D. The scenarios considered include a step reactivity insertion and a pump trip for the molten salt reactors. Overall, the results obtained so far demonstrates the potential of the developed methodology at CNL in simulating the non-water cooled concepts.

Study of Electrolyzer Materials at High Tritium Concentrations

Electrolysis cells are required to drive the combined electrolysis and catalytic exchange process used in heavy water upgrading and water detritiation. Past projects have used very robust alkaline electrolyte technology for the electrolysis cells, though recently there has been a move toward proton exchange membrane (PEM) technology. In PEM electrolysis a solid polymer electrolyte (SPE) acts as the proton conductor, separator of product gases, and insulator between electrodes. The long-term effects of highly tritiated water on these SPE materials are not fully understood. At Canadian Nuclear Laboratories (CNL), an exposure study has been undertaken wherein various commercial and proprietary SPE materials were exposed to very highly tritiated water (~1000 Ci/kg, 37 TBq/kg). Exposures were done at a typical cell operating temperature (60°C) for periods that might be expected for commercial operations. Following exposure, some samples lost sufficient integrity that they could not undergo post-exposure testing. In order to test the remaining materials’ electrolytic performance and physical properties in a nonactive laboratory, a process of decontamination that would result in no further membrane degradation needed to be developed. The successful reduction in tritium content of the samples following decontamination was verified using chemical digestion and combustion analysis. All types of commercial membranes were found to lose significant ion exchange capacity, to show reduced water absorption, and to show reduced strain before failure. Tensile testing showed almost complete degradation even at low doses. In this paper, commercial membrane data are compared with data from CNL’s tritium-compatible membranes.

Investigation of in-beam prompt and delayed neutron counting techniques for detection and characterization of special nuclear material

A neutron counting (NC) technique for in-beam active neutron interrogation has been developed and deployed at the National Research Universal (NRU) reactor of the Canadian Nuclear Laboratories (CNL). A distinguishing feature of the method is that the sample to be interrogated remains stationary during the irradiation and counting periods for the detection of prompt and delayed neutrons. The NC system permits measurements of prompt and delayed neutrons using a mono-energetic or broad-spectrum (white) thermal neutron beam. An MCNPX computational study of an NC system employing nine 3He tubes showed that a delayed neutron detection efficiency of 22% could be achieved. The presence of a sub-milligram mass of 235U could be revealed in less than ten minutes in the in-beam delayed neutron experiment incorporating a white thermal neutron beam. The technique readily differentiated between 235U and 233U isotopes by analysis of delayed neutron count rates. The lower flux of the in-beam DN experiment does not permit the trace analysis that is possible with in-core irradiation, but does permit non-destructive analysis of large samples and could prove invaluable for the initial survey of materials of unknown content and origin. The study also demonstrated that in-beam prompt neutron analysis could be the ultimate solution when the neutron source is weak, the sample is shielded, the fissile mass is small, or interrogation time is limited. Detection time was reduced to a few seconds in the examination of prompt neutrons either in single or coincidence method. This paper presents the application of an in-beam NC system using a mono-energetic and white thermal neutron beam at the NRU reactor and assesses the performance of a newly constructed portable NC system with a large neutron detection array.

Experimental measurements to evaluate the effects of fuel microstructure on fission gas release in thoria

Canadian Nuclear Laboratories (CNL) has many decades of thoria research and development experience, including fabrication, irradiation testing, and post–irradiation examination. CNL has previously shown fission gas release (FGR) to be dependent on fuel microstructure in thoria fuels. Lower density fuel with large granules surrounded by porous regions results in higher FGR. Higher and more uniform density fuel results in superior FGR results in thoria fuels with a similar irradiation history. In this paper, this effect of fuel microstructure on FGR is shown empirically using the most recent post–irradiation examination (PIE) results of CNL’s DME-176 and DME-221 thoria experiments. Previously reported PIE results of CNL’s BDL-422 and DME-221 thoria experiments are also compared.

Canadian advances in the copper–chlorine thermochemical cycle for clean hydrogen production: A focus on electrolysis

Hydrogen is a clean energy carrier that can help mitigate greenhouse gas (GHG) emissions if it is used to replace fossil fuels for power production. One way to produce hydrogen on a large scale is through the use of water splitting thermochemical cycles such as the hybrid copper chlorine (Cu–Cl) cycle. Canadian Nuclear Laboratories Ltd. (CNL) chose to develop the Cu–Cl cycle because the highest temperature required by this cycle is about 530 °C, compatible with the Canadian Super Critical Water Reactor (SCWR) or some small modular reactors (SMR). The on-going effort at CNL is to demonstrate a fully integrated Cu–Cl cycle at laboratory scale with a hydrogen production rate of 50 L/h. Some recent experimental results of the electrolysis step, one of the main steps of the cycle, are discussed in this paper.The anode reaction of CuCl oxidation was investigated using a three-electrode electrochemical cell. Half-cell experiments found that CuCl oxidation did not require noble metals as catalyst. The CuCl oxidation on carbon was found to be a mass-transfer controlled process. Hence the limiting current density increased with increasing turbulence on the electrode surface. Increasing the CuCl concentration and the solution temperature also resulted in higher limiting current densities. A current density of 0.53 A/cm2 was achieved for a 1.0 M CuCl solution at 80 °C.Single cells with electrode areas up to 100 cm2 were used to establish the operating conditions for the electrolysis step. The effects of flow rate, temperature, and current density on the cell voltage were studied. A hydrogen production rate of 50 L/h was successfully achieved at 0.4 A/cm2 in a 2.0 M CuCl solution at 80 °C. The electrolysis step is fully developed for integration in a laboratory-scale demonstration of the Cu–Cl cycle.

Hydrogen deflagrations in stratified flat layers in the large-scale vented combustion test facility

This paper examines the flame dynamics of vented deflagration in stratified hydrogen layers. It also compares the measured combustion pressure transients with 3D GOTHIC simulations to assess GOTHIC's capability to simulate the associated phenomena. The experiments were performed in the Large-Scale Vented Combustion Test Facility at the Canadian Nuclear Laboratories. The stratified layer was formed by injecting hydrogen at a high elevation at a constant flow rate. The dominant parameters for vented deflagrations in stratified layers were investigated. The experimental results show that significant overpressures are generated in stratified hydrogen–air mixtures with local high concentration even though the volume-averaged hydrogen concentration is non-flammable. The GOTHIC predictions capture the overall pressure dynamics of combustion very well, but the peak overpressures are consistently over-predicted, particularly with higher maximum hydrogen concentrations. The measured combustion overpressures are also compared with Molkov's model prediction based on a layer-averaged hydrogen concentration.

Nd isotope mapping of a tectonic window near Chalk River, Ontario: Implications for large-scale Grenvillian tectonics

Over forty new Nd isotope determinations are presented for granitoid orthogneisses from the Chalk River area, on the border between Ontario and Quebec in the SW Grenville Province. TDM model ages provide evidence for a tectonic window through the allochthonous belt into the Grenvillian parautochthon in the Chalk River area. This tectonic window constrains the location of the main ramp of the Allochthon Boundary Thrust southeast of the Chalk River area. However, the structure of the tectonic window has been complicated by down-faulting on the Ontario side of the Ottawa River, due to reactivation of late Precambrian normal faulting associated with the Ottawa–Bonnechere graben. Thus, on the Ontario side of the border, the surface expression of the tectonic window is represented by a strongly sheared tectonic duplex that underlies the main allochthon. However, the lower section of a 1 km-deep borehole drilled by Canadian Nuclear Laboratories has TDM model ages indicating penetration into the underlying parautochthon. The new data, combined with previous Nd isotope mapping and Lithoprobe seismic reflection results, point to a relatively constant thickness for the exhumed Muskoka allochthon. By showing that this exhumation occurred immediately below the Central Metasedimentary Belt boundary, this evidence strongly supports a ramp-flat thrusting model for the SW Grenville province. This model greatly reduces the likelihood of major crustal exhumation to the NW of Parry Sound domain, thus removing the theoretical basis for pervasive southeast directed orogenic collapse across the Ontario gneiss belt.

A large-scale study on the effect of ambient conditions on hydrogen recombiner-induced ignition

Hydrogen recombiners (known in the nuclear industry as passive autocatalytic recombiners-PARs), in general, can be utilized for mitigation of hydrogen in controlled areas where there is potential for hydrogen release and ventilation is not practical. Recombiners are widely implemented in the nuclear industry, however there are other applications of recombiners outside the nuclear industry that have not yet been explored practically. The most notable benefit of recombiners over conventional hydrogen mitigation measures is their passive capability, where power or operator actions are not needed for the equipment to remove hydrogen when it is present.One of most significant concerns regarding the use of hydrogen recombiners in industry is their potential to ignite hydrogen at elevated concentrations (>6 vol%). The catalyst, heated by the exothermal H2–O2 reaction, is known to be a potential ignition source to cause hydrogen burns. An experimental program utilizing a full-size PAR at the Large-Scale Vented Combustion Test Facility (LSVCTF) has been carried out by Canadian Nuclear Laboratories (CNL) to investigate and understand the behaviour of hydrogen combustion induced by a PAR on a large-scale basis. A number of parameters external to the PAR have been explored including the effect of ambient humidity (steam) and temperature. The various aspects of this investigation will be discussed in this paper and examples of results are provided.

Measurements on (233U, Th)O2 fuel by substitution experiments in the ZED-2 critical facility

This paper summarizes the results of experimental measurements on mixed-oxide (233U,Th)O2 fuel bundles in the ZED-2 heavy water critical facility of Canadian Nuclear Laboratories (CNL) in Chalk River, Canada. Pressure-tube heavy water reactors can use such fuel to extract the energy potential of thorium, which helps extend nuclear energy resources and reduces waste. Measurements were made in a series of substitution ZED-2 experiments, with (233U,Th)O2 test fuel placed into vertical channels filled with air or heavy water in a reference lattice of very low-enriched uranium fuel. Fuel was arranged to maximize the importance of the test fuel. Each fuel and coolant configuration in the substitution series was simulated in MCNP5 to verify the measurement of critical moderator height. The precision of the substitution measurements can help support the validation of reactor physics codes for evaluation of reactivity coefficients in reactors with fuel that contains oxides of thorium and thorium-derived uranium.

Construction of a thorium/actinium generator at the Canadian Nuclear Laboratories

Targeted Alpha Therapy (TAT) has demonstrated considerable promise in the treatment of a range of cancers in both preclinical and, more recently clinical research. In particular, work with the alpha-emitting radionuclide 225Ac has been effectively employed due to the relatively rapid decay cascade that leads to 4 alpha and 2 beta emissions. One limitation for TAT has been caused by access to the vital radionuclide. Traditionally, 225Ac has been sourced from thorium/actinium generators based on the alpha decay of stockpiles of 229Th. 229Th is itself the alpha-decay product from 233U. Due to proliferation issues associated with 233U, only three thorium/actinium generators have been reported in the literature, capable of supporting clinical research. This paper describes the construction and operation of a thorium/actinium radionuclide generator at the Canadian Nuclear Laboratories, capable of supporting preclinical and limited clinical research in the area of TAT. Thorium was recovered and purified by a combination of anion exchange and extraction chromatography from aged 233U stockpiles. A separation scheme for 225Ra and 225Ac has been developed, based upon the chemical composition of the thorium material to allow for regular, routine milkings capable of supplying up to 3.7 GBq (100 mCi) of radiochemically pure 225Ac annually. This routine separation is accomplished using a combination of anion exchange chromatography to separate Ac and Ra isotopes from Th and extraction chromatography employing TEVA and DGA-N resins to separate actinium from radium and breakthrough thorium.

Special Section: Selected Papers From the Ninth International Symposium on Supercritical Water-Cooled Reactors

Special Section: Selected Papers From the Ninth International Symposium on Supercritical Water-Cooled Reactors

Recent measurements on plutonium-thorium fuel by substitution experiments in the ZED-2 critical facility

This paper summarizes the results of experimental measurements recently made on heavy water moderated plutonium-thorium (Pu-Th) mixed-oxide fuel bundles in the ZED-2 critical facility of Canadian Nuclear Laboratories. Pu-Th fuel in pressure-tube Heavy water reactors can help extend nuclear energy resources by utilizing plutonium from spent fuel, and extracting the energy potential of thorium. The ZED-2 experiment was a series of substitution measurements, with (Pu,Th)O2 test fuel placed into vertical channels filled with air or heavy water in a reference lattice of very low-enriched uranium (LEU) fuel. Sensitivity to the test fuel was optimized by placement into positions of higher power. MCNP5 simulations of the fuel substitution series are within uncertainty of critical moderator height measurements. The precision of the substitution measurements can help support the validation of reactor physics codes for evaluation of reactivity coefficients in PT-HWRs with fuel that contains a mixture of thorium and plutonium oxides.

Development of a Research-Scale Thorium/Actinium Generator at the Canadian Nuclear Laboratories

The limited supply and relative cost of Actinium-225 can be prohibitive to researchers in the field of Targeted Alpha Therapy (TAT). Currently, there are only three sources (Institute for Transuranium Elements (ITU), Germany; Oak Ridge National Laboratory (ORNL), USA; and Institute of Physics and Power Engineering (IPPE), Russia) of commercially available, isotopically pure Actinium-225. Two Thorium-229/Actinium-225 generators have been developed at Canadian Nuclear Laboratories (CNL) by isolation of Thorium-229 from an aged Uranium-233 supply. The generators are capable of producing 3-6 mCi of isotopically pure Actinium-225 monthly. The Actinium-225 yielded from the generators has been used in radiolabelling experiments with novel Actinium-225 chelates and conjugated monoclonal antibodies, and has supported pre-clinical studies utilizing Actinium-225 labelled targeting vectors. Radiolabelling experiments have determined that the Actinium-225 material produced from the generators is suitable for use in preparation of pre-clinical scale TAT agents.